Field
The disclosed and claimed concept relates generally to nuclear power generation equipment and, more particularly, to a method of predicting an amount of wear that is expected to occur on the tubes of a steam generator.
Related Art
As is understood in the relevant art, pressurized water nuclear reactors employ a primary loop that includes radioactive water that flows through the reactor core and a secondary loop that receives heat from the primary loop which is used to perform mechanical work. Such heat is communicated from the primary loop to the secondary loop by employing a steam generator having a large number of tubes that are connected in fluid communication with the primary loop. The steam generator also includes a plenum within which the fluid of the secondary loop flows into contact with the exterior surfaces of the tubes of the steam generator. The steam generator typically additionally includes anti-vibration bars and other structures that resist or at least limit the vibration of the tubes within the interior of the steam generator.
While such steam generators have been generally effective for their intended purposes, they have not been without limitation. Despite the existence of the anti-vibration bars within the interior of the steam generator, the tubes of the steam generator nevertheless experience a certain level of vibration and typically vibrate against the anti-vibration bars and other structures, thus resulting in fretting wear at certain locations on the exterior surfaces of the tubes. Such wear must be monitored closely in order to avoid a situation wherein the wear would be of sufficient magnitude that the wall of a tube would be breached, which would result in undesirable nuclear contamination between the primary and secondary loops. Additionally, regulations imposed by the United States Nuclear Regulatory commission (NRC), require the tube(s) to be physically plugged when the magnitude of the wear exceeds a value of 40% of the tube wall thickness. However, for this example that level at which plugging is required is taken as 100% of the tube wall thickness for illustration purposes only. As such, the tubes of the steam generator are periodically inspected through the use of an eddy current sensor that is received in the tubes and that is advanced along the tubes in a known fashion while signals from the sensor are detected and recorded. The signals from the eddy current sensor are usable to determine, for instance, a depth of wear on the exterior of a tube at a location thereon.
By knowing the thickness of the tube wall, the wear analysis that has heretofore been employed would rely upon a straight line depth of wear analysis to predict wear on the tube. For instance, if at a given previous time it had been determined that 20% of the wall thickness had been worn away at a specific location, and that at a current time 60% of the wall thickness had been worn away at the specific location, the analysis would conclude that during the time interval between the two times at which measurements were taken, an additional 40% of the tube wall thickness had been worn away. Employing the same straight line depth of wear analysis, this methodology would predict that at a future time after another time interval equal to the previous time interval, another 40% of the wall thickness would be expected to worn away. In the present example, such wear would include the 60% wall thickness worn away at the current time plus an additional predicted 40% wall thickness worn away at the future time, which would equal 100% of the wall thickness being worn away at the future time, and this indicates an undesirable breach of the tube wall.
Since such inspections typically occur during refueling of a nuclear reactor and thus are at regular time intervals, it was possible, using the aforementioned analysis, to determine whether certain tubes should be plugged prior to the steam generator and the reactor being placed back into service. It is known, however, that the plugging of a tube of a steam generator is undesirable because it reduces the power output that can be obtained from a nuclear reactor. Improvements thus would be desirable.